1. Field of the Invention
The present invention generally relates to zirconium based alloys and their use in the formation of strips and tubing for nuclear reactor fuel assemblies. In particular, the invention relates to improving in-reactor corrosion and/or in-reactor creep of Zr—Nb based alloys by a final heat treatment. More particularly, the invention relates to improving, i.e. reducing, in-reactor corrosion by limiting or restricting the amount of iron in the alloy composition.
2. Description of the Prior Art
In the development of nuclear reactors, such as pressurized water reactors and boiling water reactors, fuel designs impose significantly increased demands on all of the fuel components, such as cladding, grids, guide tubes, and the like. Such components are conventionally fabricated from zirconium-based alloys commercially known as ZIRLO, corrosion resistant alloys that contain about 0.5-2.0 wt. % Nb; 0.9-1.5 wt. % Sn; and 0.09-0.11 wt. % of a third alloying element selected from Mo, V, Fe, Cr, Cu, Ni, or W, with the remainder Zr, as taught in U.S. Pat. No. 4,649,023 (Sabol et al.). That patent also taught compositions containing up to about 0.25 wt. % of the third alloying element, but preferably about 0.1 wt. %. Sabol et al., in “Development of a Cladding Alloy for High Burnup” Zirconium in the Nuclear Industry: Eighth International Symposium, L. F. Van Swan and C. M. Eucken, Eds., American Society for Testing and Materials, ASTM STP 1023, Philadelphia, 1989. pp. 227-244, reported improved properties of corrosion resistance for ZIRLO (0.99 wt. % Nb, 0.96 wt. % Sn, 0.10 wt. % Fe, remainder primarily zirconium) relative to Zircaloy-4.
There have been increased demands on such nuclear core components, in the form of longer required residence times and higher coolant temperatures, both of which increase alloy corrosion. These increased demands have prompted the development of alloys that have improved corrosion and hydriding resistance, as well as adequate fabrication and mechanical properties. Further prior art in this area include U.S. Pat. Nos. 5,940,464; 6,514,360 (Mardon et al. and Jeong et al.) and Reexamination Certificate U.S. Pat. No. 5,940,464 C1 (both Mardon et al.), and the technical paper “Advanced Cladding Material for PWR Application: AXIOM™”, Pan et al., Proceedings of 2010 LWR Fuel Performance/Top Fuel/WRFPM, Orlando, Fla. Sep. 26-29, 2010 (“technical paper”),
Mardon et al. taught zirconium alloy tubes for forming the whole or outer portion of a nuclear fuel cladding or assembly guide tube having a low tin composition: 0.8-1.8 wt. % Nb; 0.2-0.6 wt. % Sn, 0.02-0.4 wt. % Fe, with a carbon content of 30-180 ppm, a silicon content of 10-120 ppm and an oxygen content of 600-1800 ppm, with the remainder Zr. Jeong et al. taught a niobium-containing zirconium alloy for high burn-up nuclear fuel application containing Nb, Sn, Fe, Cr, Zr with optional addition of Cu. The Pan et al. “technical paper” includes alloys listed as X1, X4, X5, X5A, but only generally describes the actual composition weight percentages. Pan et al. reports tensile strength, elongation and creep test data, and shows micrographs and in-reactor corrosion and oxide thickness data.
Aqueous corrosion in zirconium alloys is a complex, multi-step process. Corrosion of the alloys in reactors is further complicated by the presence of an intense radiation field which may affect each step in the corrosion process. In the early stages of oxidation, a thin compact black oxide film develops that is protective and retards further oxidation. This dense layer of zirconia exhibits a tetragonal crystal structure which is normally stable at high pressure and temperature. As the oxidation proceeds, the compressive stresses in the oxide layer cannot be counterbalanced by the tensile stresses in the metallic substrate and the oxide undergoes a transition. Once this transition has occurred, only a portion of the oxide layer remains protective. The dense oxide layer is then renewed below the transformed oxide. A new dense oxide layer grows underneath the porous oxide. Corrosion in zirconium alloys is characterized by this repetitive process of growth and transition. Eventually, the process results in a relatively thick outer layer of non-protective, porous oxide. There have been a wide variety of studies on corrosion processes in zirconium alloys. These studies range from field measurements of oxide thickness on irradiated fuel rod cladding to detailed micro-characterization of oxides formed on zirconium alloys under well-controlled laboratory conditions. However, the in-reactor corrosion of zirconium alloys is a complicated, multi-parameter process. No single theory has yet completely defined it.
Corrosion is accelerated in the presence of lithium hydroxide and pressurized water reactor (PWR) coolant contains lithium. Thus, it is desired to limit or preclude acceleration of corrosion due to concentration of lithium in the oxide layer. U.S. Pat. Nos. 5,112,573 and 5,230,758 (both Foster et al.) taught an improved ZIRLO composition that was economically produced and provided an easily controlled composition while maintaining corrosion resistance similar to previous ZIRLO compositions. It contained 0.5-2.0 wt. % Nb; 0.7-1.5 wt. % Sn; 0.07-0.14 wt. % Fe and 0.03-0.14 wt. % of at least one of Ni and Cr, with the remainder Zr. This alloy had a 520° C. high temperature steam weight gain at 15 days of no greater than 633 mg/dm2. U.S. Pat. No. 4,938,920 to Garzarolli teaches a composition having 0-1 wt. % Nb; 0-0.8 wt. % Sn, and at least two metals selected from iron, chromium and vanadium. However, in Garzarolli when niobium is present, tin is absent and vice versa. Therefore, does not disclose an alloy that had both niobium and tin.
Sabol et al, in “In-Reactor Corrosion Performance of ZIRLO and Zircaloy-4,” Zirconium in the Nuclear Industry: Tenth International Symposium, A. M. Garde and E. R. Bradley Eds., American Society for Testing and Materials, ASTM STP 1245, Philadelphia 1994, pp. 724-744, demonstrated that, in addition to improved corrosion performance, ZIRLO material also has greater dimensional stability (specifically, irradiation creep and irradiation growth) than Zircaloy-4. More recently, U.S. Pat. No. 5,560,790 (Nikulina et al.) taught zirconium-based materials having high tin contents where the microstructure contained Zr—Fe—Nb particles. The alloy composition contained: 0.5-1.5 wt. % Nb; 0.9-1.5 wt. % Sn; 0.3-0.6 wt. % Fe, with minor amounts of Cr, C, O and Si, with the remainder Zr.
While these modified zirconium based compositions are described as providing improved corrosion resistance as well as improved fabrication properties, economics have driven the operation of nuclear power plants to higher coolant temperatures, higher burn-ups, higher concentrations of lithium in the coolant, longer cycles, and longer in-core residence times that have resulted in increased corrosion for the cladding. Continuation of this trend, as burn-ups approach and exceed 70,000 MWd/MTU, will require further improvement in the corrosion properties of zirconium based alloys. The alloys of this invention provide such corrosion resistance.
Another potential mechanism for increasing corrosion resistance is through the method of forming the alloy itself. In general, to form alloy elements into a tubing or strip, ingots are conventionally vacuum melted and beta quenched, and thereafter formed into an alloy through various reductions, intermediate anneals, and final anneals, wherein the intermediate anneal temperature is typically above 1105° F. for at least one of the intermediate anneals. In U.S. Pat. No. 4,649,023 to Sabol et al., the ingots are extruded into a tube after being beta quenched and beta annealed, and thereafter alternatively cold worked in a pilger mill, and intermediately annealed at least three times. While a broad range of intermediate anneal temperatures are disclosed, the first intermediate anneal temperature is preferably 1112° F., followed by a later intermediate anneal temperature of 1076° F. The beta annealing step preferably uses temperatures of about 1750° F. Foster et al., in U.S. Pat. No. 5,230,758, determined the formability and steam corrosion for intermediate anneal temperatures of 1100° F., 1250° F., and 1350° F. An increase in intermediate anneal temperature is associated with an increase in both formability and corrosion resistance. U.S. Pat. No. 5,887,045 to Mardon et al. discloses an alloy forming method having at least two intermediate annealing steps carried out between 1184° to 1400° F.
The prior art for corrosion improvement as summarized above involves alloying element additions and different intermediate anneal temperatures, but, does not teach a final anneal heat treatment temperature. Rudling et al., in, “Corrosion Performance of Zircaloy-2 and Zircaloy-4 PWR Fuel Cladding,” Zirconium in the Nuclear Industry: Eight International Symposium, ASTM STP 1023, L. F. Van Swam and C. M. Eucken, eds. American Society for Testing and Materials, Philadelphia, 1989, pp. 213-226, reported that Zr-4 fuel rod cladding fabricated from the same ingot with final heat treatments of stress-relieved (SRA) and fully recrystallized (RXA) exhibited similar oxide thickness corrosion (see Table 1).
TABLE 1Post irradiation oxide thickness of Zr-4cladding after 1-cycle of irradiation.Final Heat4 Rod Average of the MaximumTreatmentOxide Thickness (μm)SRA12 +/− 1RXA10 +/− 1
Foster et al., in U.S. Pat. No. 5,125,985, presents a method of controlling creep by use of a final area reduction and intermediate anneal temperature. A decrease in final area reduction decreases creep, and an increase in intermediate anneal temperature decreases creep. In different applications, the in-reactor creep may be more important than in-reactor corrosion. One such example is fuel rods containing fuel pellets coated with ZrB2. ZrB2 is a neutron absorber. When neutrons are absorbed, He gas is released which increases the rod internal pressure. In this case, creep resistant cladding is necessary so that the fuel/cladding gap remains closed. A closed fuel/cladding gap ensures the fuel temperatures do not increase due to the formation of a He gas gap between the fuel and cladding. In accordance with the present invention, either the cladding corrosion or the cladding in-reactor creep may be improved by the final heat treatment.
A further issue in nuclear reactors is corrosion of welds utilized in a nuclear fuel assembly. In a typical fuel rod, nuclear fuel pellets are placed within the cladding, which is enclosed by end caps on either end of the cladding, such that the end caps are welded to the cladding. The weld connecting the end caps to the cladding, however, generally exhibits corrosion to an even greater extent than the cladding itself, usually by a factor of two over non-welded metal. Rapid corrosion of the weld creates an even greater safety risk than the corrosion of non-welded material, and its protection has generally not been addressed. In addition, grids have many welds and the structural integrity depends on adequate weld corrosion resistance.
Thus, there is a need, even in this later stage of nuclear power development, for novel zirconium cladding alloys that exhibit improved corrosion resistance and improved in-reactor irradiation creep resistance over known alloys in the field, and improved welds for holding end caps on claddings and for joining grid straps that likewise exhibit increased corrosion resistance. Accordingly, an object of the present invention is to provide Zr—Nb alloys with improved corrosion resistance and/or improved in-reactor irradiation creep resistance through the selection of a specific final heat treatment and restriction of the amount of iron in the material chemical composition.